The following pertains to the nuclear reactor arts, nuclear power arts, nuclear reactor safety arts, and related arts.
Existing nuclear power plants are typically light water thermal nuclear reactors of the boiling water reactor (BWR) or pressurized water reactor (PWR) designs. In such a reactor, a nuclear reactor core comprising fissile material (typically a uranium compound such as UO2 enriched in fissile 235U) is disposed in coolant (purified water) contained at an operational pressure and temperature in a reactor pressure vessel. A nuclear chain reaction involving fission of the fissile 235U generates heat in the nuclear reactor core which is transferred to the coolant. In a BWR design, the heat directly converts coolant to steam, that is used to drive a turbine, while in a PWR design, the primary coolant remains in a liquid state (e.g. subcooled) and converts secondary coolant to steam in a steam generator—this secondary coolant steam is then used to drive the turbine. In conventional PWR designs the steam generator is a component external from the reactor pressure vessel that contains the nuclear reactor. In so-called integral PWR designs, the steam generator is located integrally inside the reactor pressure vessel with (typically above) the nuclear reactor core. The integral PWR design advantageously reduces the number of large-diameter vessel penetrations carrying primary coolant. In some contemplated PWR designs, these vessel penetrations may be limited to primary coolant let-down and make-up lines, and emergency core cooling (ECC) connections.
Safe operation of a nuclear reactor includes providing protection against radiological release to the environment. To this end, it is known to surround the nuclear reactor with a radiological containment structure typically constructed of steel and/or steel-reinforced concrete, and to implement safety systems, with redundancy, to remediate events in which reactor operation moves outside of a design envelope. One class of events is a loss of coolant accident (LOCA), in which reactor coolant escapes from a reactor pressure vessel break or, more commonly, from a break in a large-diameter pipe that connects with the reactor pressure vessel at a vessel penetration. A LOCA break which occurs between the vessel penetration and a closest pipe valve is particularly problematic, since reactor coolant loss from such a break continues even after the pipe valve is closed. A known solution is to provide an integral isolation valve (IIV) at the vessel penetration. An IIV comprises a valve built into a flange that connects with the pressure vessel. Since the IIV is integrated directly into the vessel penetration, closing the IIV ensures stoppage of reactor coolant loss at the LOCA break. Some IIV designs are disclosed in Malloy III et al., U.S. Pub. No. 2013/0272478 A1 published Oct. 7, 2013, and in Malloy III et al., U.S. Pub. No. 2013/0294567 A1 published Nov. 7, 2013.
An IIV protecting the make-up line can be constructed as a check valve (or two check valves in series to provide redundancy) that is oriented to prevent flow of coolant out of the nuclear reactor. A check valve(s) is suitable because the make-up coolant flows into the reactor vessel. On the other hand, an IIV protecting the let-down line is more difficult. This is because the let-down coolant flows out of the reactor, so that a check valve design would prevent normal let-down operation. Accordingly, IIV designs for the let-down line typically employ a motor-operated valve.